Effect of Reactor Radiation on the Thermal Conductivity of Graphite-based Dispersion Fuel
- Argonne National Laboratory: 9700 S Cass Ave, Lemont, Illinois, 60439 (United States)
The Fukushima Daiichi nuclear accident in 2011 stimulated materials development efforts to improve the current fuel design for Light Water Reactors (LWRs). From fuel element to fuel cladding, new concepts in advanced nuclear fuel development have been proposed, studied, and tested. The new fuel systems are expected to have better resistance to damage, melting, and chemical reaction during accident scenarios. To test the performance and accident tolerance of the advanced fuel, it is highly preferable to simulate off-normal transients in addition to standard in-core fuel performance tests (available at High Flux Isotope Reactor at Oak Ridge National Laboratory and Advanced Test Reactor at Idaho National Laboratory). The Transient Reactor Test Facility (TREAT) is now being prepared to restart its operation, performing transient testing to assist advanced fuel development. TREAT was operated from 1959 to 1994, and was extensively used to generate thousands of reactor transients for different programs. The reactor has not operated since 1994. The original fuel used at TREAT was graphite-based dispersion fuel with highly-enriched uranium (HEU, 93.1% enriched uranium). For nonproliferation purposes, a fuel conversion program, supported by U.S. Department of Energy, National Nuclear Safety Administration (NNSA), Office of Material Management and Minimization (NA-23) Reactor Conversion Program, is underway to develop a replacement core in which the UO{sub 2} contains only low-enriched uranium (LEU). The preliminary LEU core design has shown a much lower carbon/uranium atomic ratio and enrichment than in the HEU core design. To achieve performance similar to the HEU core, the prospective LEU core will contain a much higher density of UO{sub 2} particles. As a result, a higher level of radiation damage, particularly from fission fragments emerging from the UO{sub 2} particles, can be expected dependent on the UO{sub 2} particle size within the fuel. To support LEU fuel design, the present study investigates the effect of reactor radiations on the thermal conductivity of graphite-based dispersion fuel. Both neutron- and fission-radiation induced degradations of the fuel are modeled, and validated based on the available experimental data. In this study, the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel was analyzed. A semi-empirical method to model the relative thermal conductivity with reactor radiations was proposed and validated. The proposed method can be further applied to predict the changes in the thermal conductivity of the prospective LEU TREAT fuel during long-term operation. (authors)
- OSTI ID:
- 22992126
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 114, Issue 1; Conference: Annual Meeting of the American Nuclear Society. Embedded topical meeting 'Nuclear fuels and structural material for the next generation nuclear reactors', New Orleans, LA (United States), 12-16 Jun 2016; Other Information: Country of input: France; 14 refs.; Available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 United States; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
36 MATERIALS SCIENCE
FISSION
GRAPHITE
HFIR REACTOR
HIGHLY ENRICHED URANIUM
IDAHO NATIONAL LABORATORY
MELTING
MINIMIZATION
NUCLEAR FUELS
RADIATION ACCIDENTS
RADIATION EFFECTS
RADIATION PROTECTION
REACTOR ACCIDENTS
THERMAL CONDUCTIVITY
TREAT REACTOR
URANIUM DIOXIDE